During neutron irradiation of ceramic breeders, not only does the transmutation of Li to T and He take place but other isotopes are formed as well, and the impurities contained will all contribute to the induc­tion of radioactivity from exposure to the neutron irradiation environment. This enables assessment of the feasibility of recycling ceramic breeder material from blanket components having reached EOL con­ditions. Recycling options for ceramic breeder material not only avoid large waste volumes requiring long-term storage but also contribute to resource effi­ciency of valuable constituents such as lithium. From a ceramic breeder perspective, Li2O, Li2TiO3, and Li4SiO4 are more attractive than Li2ZrO3 due to their long-term activation characteristics.

Knitter et a/.187 assessed contact dose rates of different Li4SiO4 materials for high radiation levels expected from a fusion power reactor for 1 FPY. As an example, Figure 64 displays the contact dose rate versus time after shutdown for pure Li4SiO4, which specifically results from the production of 28Al, 24Na, 7Be, and 26Al. Due to the activity of 7Be, the recycling limit for remote handling (10mSvh~) and the
hands-on limit (10muSvh-1) are reached after <1 and <4 years, respectively. The activation of as manufactured Li4SiO4 appears to be strongly influenced by impurities, of which Co and Pt appear to be the most important.187 Knitter et a/. also simulated remelting of lithium depleted Li4SiO4

experimentally.188

Pebble manufacturing routes by wet processes require different considerations for reprocessing.196 In the powder preparation stage the powders need to be produced from pre-existing pebbles by dissolution and precipitation steps; see, for example, Tsuchiya eta/.184

The activation and transmutation of tungsten as a PFM is a critical issue, particularly concerning long-term storage and recycling times. Different studies on activation issues have been performed. These comprise the analysis of cross-sections for high-energy neutrons,1 , 2 studies on the helium — cooled lithium lead divertor for DEMO,113 the inertial fusion devices,109 other benchmark experiments,114,115 and modeling issues, for example, on the self-shielding ability of tungsten.116,117

Furthermore, it was shown that the long-term activation behavior is dominated by activation products of the assumed material impurities while the short­term behavior is due to the activation of the stable W isotopes.113 For a short period of a few weeks, the latter causes a huge amount ofdecay-induced afterheat that has to be removed by continued active cooling.67 On the other hand, the accumulation of the highly radioactive transmutation product 186mRe was deter­mined to be most critical, limiting the component lifetime to a maximum of 5 fpy when using pure W or to 2 fpy when using Re-doped W before the limits for storage by shallow land burial could be exceeded.109 The dose rate limits for recycling after different applications are expected to be reached within 5 years115 to 50 years113 of storage or up to 75 years after end of plant life.118 Fischer etal.113 take a limit of 100 mSv h-1 for remote handling into account, which might be a problem at the times when maintenance operations would be in progress. Taylor and Pampin118 give a value of 20mSvh-1 as the limit for allowing tungsten to be categorized as a recyclable material. The hands-on limit for tungsten should be achieved after about 200 years.115

Besides waste management, tungsten has also been investigated and evaluated according to characteristic radiological hazards that might occur when using it as PFM in tokamak fusion reactors. It was found that the tritium permeation into tungsten does not, in contrast to CFC, appear to be a major problem. However, due to neutron activation, the mobilization of activation products, for example, by forming volatile oxide spe­cies in the presence of steam and air, has to be limited by establishing shutdown requirements to avoid melting of tungsten in case of an accident. The poten­tial exposure from mobilized activation products from the tungsten divertor may be modified by vary­ing the operating conditions of fusion power and change-out time as well as the thickness of the divertor armor. The dose can be reduced by selecting shorter change-out times. However, the total life­cycle waste volume will be increased accordingly. A thinner divertor will produce less mobilized acti­vation products while suffering a more restrictive shutdown requirement.11

When the eV or slightly more energetic ions of hydrogen atoms impact the graphite plasma-facing surface, they will chemically react with the surface. The form this reaction takes is quite dependent on the temperature of the graphite, and as the majority of the ions are reacting over nanometer lengths, the result is the production of an amorphous film of hydrogenated carbon suit at the near-surface layer. An analytical expression for this thermally activated erosion process has been put forward by Roth and Garcia-Rosales,45 with further development by Mech46 and Roth.47 It is noted that in the ion energies between 10 and 20 eV, there is still an active debate regarding the synergy ofthe ion impact and the chemical erosion.

Temperature (°C)

The combination of energetic damage plus chemical reaction, which is sometimes referred to as ‘chemi­cal sputtering,’ is discussed by Jacob and Roth48 and others.

For low(RT) and intermediate temperatures, from 400 to 1000 °C (Figure 21), the volatilization of car­bon atoms by energetic plasma ions becomes impor­tant. As seen in the upper curve of Figure 21, helium does not have a chemical erosion component of its sputter yield. In the currently operating machines, the two major contributors to chemical erosion are the ions of hydrogen and oxygen. The typical chemi­cal species that evolve from the surface as measured by residual gas analysis49 and optical emission50 are hydrocarbons, carbon monoxide, and carbon dioxide.

The interaction of hydrogen with graphite appears to be highly dependent on the ion species, on mate­rial temperature, and on the perfection and type of the graphite. This is illustrated in Figure 22, which shows typical bell-shaped thermally acti­vated erosion yield curves for hydrogen and deute­rium ions on graphite. The shape of the yield curve is influenced by the competition for hydrogenation from the sp2 and sp3 hybridization states.51,52 Hydrogen ions incident to the surface are slowed down and pref­erentially attach to sp2 carbon atoms (such as graph­ite edge plane atoms) forming sp3 CH3 complexes. Above approximately 400 K, these CH3 complexes can be released, thus returning the structure to the sp2 state. It is important to note that this phenomenon will only happen in the presence of simultaneous ion damage. It will not occur simply due to a thermal process. This step leads to chemical erosion products (a host of erosion species are possible). The ability of hydrogen to continue to be bonded to carbon drops as the temperature goes up. If there are no CH or CH2 precursors on the surface, then no volatile CH3 or CH4 complexes can be formed, and thus there is no chemical erosion. This balance yields a maximum erosion rate, which for undamaged pyrolytic graphite resides at ^280-600 °C.53 It is noted that more recent work by Balden54 has determined the maximum to be in the range of 872-1222 °C. This mechanism was first elucidated by Horn55 and Wittmann.56 The rate of formation of CH2, CH3, and complex hydrocar­bons from atomic hydrogen in well-graphitized mate­rial is fairly low unless the material is altered (damaged) in the near-surface layer. For preirra­diated pyrolytic graphite (i. e., damaged graphite, meaning that a carbon atom has been removed from its lattice position, thus increasing the available sp2 sites) preirradiated by high-energy D+ of H+ ions, the total erosion yield following exposure to low-energy hydrogen increases dramatically. This is illustrated in the upper curves of Figure 22 that show more than an order of magnitude increase in erosion yield over
the undamaged case. This increased carbon loss has been attributed to the creation of active sites for Ho attachment.57’58 This structurally dependent mech­anism is supported by the data of Phillips et al.,59 which shows a factor of two difference in erosion yield between high — and low-quality pyrolytic graphite.

JET has completed in 2011 a large enhancement programme that includes, among other things, the installation of a beryllium wall and a tungsten divertor.

An overview of the status of the JET ITER-like wall project is presented in Matthews et at.119 The material combination chosen for the wall and the divertor is that chosen for the DT phase of ITER and experiments in JET with the new wall configura­tion will provide the first fully representative test of material migration, material mixing, and consequent tritium retention under ITER relevant conditions.180 Equally important is the opportunity to develop fully integrated scenarios and control schemes for protect­ing the wall. The project will therefore provide essen­tial information for interpreting material behavior in ITER and a sound technical basis for guiding the development of ITER scenarios.

The design layout, the main engineering challenges, and the operational limits of the JET ITER-like wall are discussed elsewhere (see, e. g., Nunes et at.,181 Thompson et at.,182 Riccardo,183 and Riccardo et at.184). Figure 17 describes the design layout and the planned material layout. It must be noted that the existing JET wall relies on a series of discrete poloidal limiters whereas at the moment ITER relies on a plasma conforming wall. The elec­trical resistivity of Be « 0.08 p. Q m is more than a 100 times lower than that of CFC («10 p. Q m). Therefore, after replacing the CFC tiles in JET, the mechanical loads due to eddy currents associated with disrup­tions, which were negligible in the case of carbon tiles, have become dominant for Be tiles and this

Beryllium W-coated CFC Inconel + 8mm Be Bulk W

Saddle c

Upper dump plate

Mus

Restraint ring protections

Inner wall cladding

Poloidal limiters

Inner wall g

Inner wall guard limiters

IW cladding for pellets

Divertor

(b)

Figure 17 (a) View of the present Joint European Torus main chamber with indications of how the carbon-fiber composite tiles will be replaced (reproduced with permission from Riccardo, V. J. Nucl. Mater. 2009, 390-391, 895-899). (b) Cross-section with allocations of materials. Image courtesy of EFDA-JET.

has posed a significant engineering challenge. Calcu­lations for the main limiter tile types clearly show that a tile of the size of the existing CFC tiles would give unacceptable eddy torques, leading to the inves­tigation of several slicing options.181 The chosen design has vertical slices with a large central block and one to three side slices, depending on the toroidal extent of the tile assembly, supported on a carrier via pins (see Figure 18). The design is defined by the balance between conflicting requirements ofeddy cur­rents (avoidance of large low resistance loops) and power handling (minimum number of vertical cuts to be shadowed). The problems associated with the design of the JET beryllium tiles (power handling capacity and disruption induced eddy currents) are discussed in detail elsewhere (see, e. g., Thompson eta/.182).

The installation of the new ITER-like wall and the NB enhancement has been completed by the mid of 2011 and operation is now restarting to provide important information for ITER.

4.21.3.3.1 Scoping by bulk immersion tests

Exploration of new candidates having superior per­formance compared to CaO and AlN was carried out using static immersion tests. Figure 4 shows the mass loss of various insulator ceramics due to exposure to static Li at high temperatures. As predicted by ther­modynamics, Er2O3 and Y2O3 showed stability supe­rior to that of CaO.2,15 For these materials, formation of LiXO2 (X = Er or Y) during exposure to Li was reported,17,18 although the impact of these changes on the coating properties remains to be assessed. In particular, the effects of Li flow on the stability of the corrosion products on the surfaces will be the key issue.

4.21.3.3.2 In situ coating with Er2O3

Based on the experience with CaO, in situ coating with Er2O3 and Y2O3 was attempted on V-4Cr-4Ti by doping Er or Y in Li and O into V-4Cr-4Ti. Coating with Y2O3 was shown to be difficult, proba­bly because there was almost no solubility of Y in Li.

On the other hand, formation of Er2O3 was con- firmed.19 Because the solubility of Er in Li is much lower than that of CaO, the stability of the coating, once formed, is much higher compared with the CaO in situ coating. The cross-section of the coating with compositional profile and coating thickness with time and temperature are shown in Figures 5 and 6, respectively.

In the effort to optimize the precharging condition of oxygen, the microstructural process for restoring oxygen in vanadium alloy substrate was clarified.2 Figure 7 shows the depth profile of hardness before and after oxygen charging, after heat treatment and

(b)

(c)

„ > [200] 1000 [020]

V — 4Cr -4Ti (NIFS-HEAT-2)

T-o "D і. ra

■W w

МіАіІаігіІіЦіїІцКИНіЦ

(a)

0 50 100 150 200 250 Depth from surface (pm)

300

Figure 7 Depth distribution of hardness and transmission electron microscope microstructure near the surfaces in the V-4Cr-4Ti substrate. (a) Before oxidation, (b) after oxidation, (c) after annealing, and (d) after in situ coating (coating was removed). Reproduced from Yao, Z.; Suzuki, A.; Muroga, T.; Yeliseyeva, O. I.; Nagasaka, T. Fusion Eng. Des. 2006, 81, 951-956, with permission from Elsevier.

after in situ coating, together with the transmission electron microscope (TEM) microstructures near the surfaces. The hardness is known to follow the approx­imate level of O in V-4Cr-4Ti.21 After oxidation for 6 h at 700 °C, the surface was covered with a complex oxide layer. After the subsequent heat treatment for 16 h at 700 °C, the matrix was composed of a high density of needle-shaped Ti—O (mostly TiO2) precipitates oriented in the (100) directions (net structure). This structure was most prominently observed after annealing at 700 °C. This is consistent with the results of the precipitation study, which showed that, although Ti interacts with impurity O already at ^200 °C, Ti-O precipitates start to form at ^700 °C in V-4Cr-4Ti alloys.22 Figure 7 also shows that the net structure near the surface disappeared after exposure to Li for 100 h at 700 °C because of the loss of oxygen. A recent study showed in situ healing capabilities with Er2O3, but further optimization of the process is required to obtain a reliable healing function.23

The mechanism of the net structure formation and supply of oxygen for the coating was elucidated using a kinetic model. The model successfully explained the experimental trends.24

In the 1970s, alternative fission fuel technologies had been developed based on packing of spheres to reduce the problems associated with excessive swelling and fragmentation of pellets.32 One of the early pebble-bed blanket designs was developed by Dalle Donne and coworkers6,7,9 at Forschungszentrum Karlsruhe (FZK), now called Karlsruhe Institute of Technology (KIT), Germany. In this concept, breeder ceramic and neutron multiplier were both shaped as small spheres or pebbles and arranged in a so-called mixed bed (see Figure 4).

The concept was based on small (0.1—0.2 mm diam­eter) pebbles of Li4SiO4 and a binary mixture of beryllium pebbles (0.1—0.2 mm and 1.5—2.3 mm diam­eter) (see Figure 5), taken from the extraction of tritium in ceramics 7 (EXOTIC-7) irradiation proj­ect.33 It was found that the compatibility of Li4SiO4 and beryllium was drastically reduced under neutron — irradiation conditions.34 This initiated the separation of breeder and neutron multiplier in different pebble beds in further blanket design evolution.

Extensive R&D on breeder pebbles has also been performed by the Canadian Atomic Energy of Canada Limited (AECL), where in particular Li2ZrO3 and Li2TiO3 were developed.9,35

With growing insight into thermodynamics and the experimental results obtained from neutron — irradiation testing, the European breeder out of tube (BOT) concept evolved into the helium-cooled pebble-bed (HCPB) concept,7 in view of preparing a test module program for ITER. This concept evolved further in Europe within the scope of the Power Plant Conceptual Study.15,16 The key features of this early HCPB concept are given in Figure 6. It consists of

Figure 6 The Helium Cooled Pebble Bed (HCPB) blanket concept from the European Power Plant Conceptual Study (PPCS), model B. Reproduced from EFDA, A. Conceptual Study of Commercial Fusion Power Plants; Final Report of the European Fusion Power Plant Conceptual Study (PPCS); Report EFDA-RP-RE-5.0; 2005.

Ceramic

From caps and grid

To FW

To caps and grid

From FW

Bypass

He inlet

He coolant

Figure 7 Evolved helium-cooled pebble-bed concept. Reproduced from Poitevin. Y; et al. Fusion Eng. Des. 2010, 85, 2340-2347.

alternating beds of ceramic breeder and beryllium pebbles, perpendicular to the plasma-facing wall, between flat coolant plates of Eurofer-97, a so-called reduced activation steel based on conventional 9Cr steels.3637

In this study, the blanket box is considered a con­sumable component, with (1) the maximum irradia­tion damage of primary wall structures set at 150 dpa (about 5 FPY (full power year)) and probably the limiting factor of the box lifetime; and (2) the burnup of the ceramic breeder and swelling of the beryllium neutron multiplier depending on the design.38

Further evolution of the HCPB line in Europe concentrated on the strategies for the ITER TBM, as explained by Poitevin and coworkers.17,39-41

The internal structure of the blanket is given in Figure 7. All structures contain dense patterns of cooling channels, with beds of Be and ceramic breeder in the form of near-spherical particles (0 0.25-0.63 mm for Li4SiO4, 0 1 mm for alternative breeder Li2TiO3, and 0 1 mm for beryllium) separated by cooled steel plates and bed heights sufficiently low (about 10 mm) to conduct heat to the cooling plates without exceeding material temperature limits. Tritium is removed from the pebble beds by a slow purge flow of helium at near­atmospheric pressure, with hydrogen (typically

0. 1 vol%) and defined levels of other constituents such as H2O, and so on for optimized integral performance.

Some alternative concepts were explored such as those using a 9-Cr steel variant with higher

He out

Neutron multiplier
pebble
(Be, beryllide)

Tritium breeder pebble (Li2TiO3)

Figure 8 Schematic of water-cooled ceramic breeder concept developed in Japan. Reproduced from Akiba, M.; Enoeda, M.; Tsuru, D.; etal. Fusion Eng. Des. 2009, 84, 329-332.

temperature resistance, see Hermsmeyer et a/.,18 or SiC-based composite structure.2

The Japanese designed water-cooled solid breeder blanket consists of two submodules (see Figure 8). One submodule consists of a module box, tritium breeding pebbles, neutron multiplier pebbles, and cooling panels. The tritium breeding pebbles and neutron multiplier pebbles are separated by the cool­ing panels, and the beds are oriented parallel to the plasma-facing wall. The module box and the cooling panels are made of reduced activation ferritic — martensitic steel, F82H.36,42 For the tritium breeding pebbles, Li2TiO3 is selected as the primary candidate material. These pebbles are about 0.2-2 mm in diam­eter, with either a monodisperse or binary size distribution.

Among the pebble-bed concepts, there are also designs for a low-pressure water-cooled ITER driver blanket (see, e. g., Lorenzetto et a/.10) and the ITER 1998 design document by Ioki and coworkers.11, Nardi et a/.13 developed ideas for a driver blanket for the reduced size ITER-FEAT. Recent work reported by Ihli et a/.21 also included mixed bed options for DEMO blankets, as shown in Figure 9.

Figure 9 Breeder inside tube concept with ceramic pebbles. Reproduced from Ihli, T.; Basu, T. K.;

Giancarli, L. M.; etal. FusionEng. Des. 2008, 83, 912-919.

Tungsten is another of the plasma-facing materials, described in Chapter 4.17, Tungsten as a Plasma­Facing Material. Like carbon, it will not be a vacuum barrier. Thus, permeation through the tungsten will not lead to tritium release directly into the environ­ment. It can lead to tritium permeation into the coolant through the coolant tubes inside the tungsten facing materials. Permeation will also affect the tri­tium inventory of the fusion device. Tungsten has excellent thermal properties with a very high melting point of 3683 K. The problem that tungsten presents to the tokamak designer is the deleterious radiation losses if tungsten is present in the plasma. Fortu­nately, the energy threshold for sputtering by
hydrogen ions is quite high, 700 eV for tritium.52 For that reason, tungsten will be used primarily in the divertor region where the energy of the impacting particles can be limited.

There are a limited number of reports on the diffusivity of hydrogen isotopes in tungsten. Frauen — felder53 measured the rate of hydrogen outgassing from saturated rolled sheet samples at temperatures over the wide range 1200-2400 K. His material was 99.95% pure tungsten. Zakahrov and Sharapov54 used 99.99% pure tungsten samples in their permeation techniques to determine the hydrogen diffusivity over a limited temperature range of 900-1060 K. In the Benamati et al55 experiments using tungsten con­taining 5% rhenium, a gaseous permeation technique was also used. These experiments were performed over a very limited temperature range of 850-885 K. Reported diffusivities are shown in Figure 5. There are a couple of reasons why the diffusion coefficient reported by Frauenfelder53 is widely accepted as most correct. The first of these reasons is the wide temperature range over which experiments were per­formed. The second reason is that the experiments were performed at a temperature above that where trapping typically occurs. It can be seen in Figure 5 that Zakahrov’s54 diffusivity agrees quite well with Frauenfelder’s at the highest temperatures, but falls below his values at lower temperatures, where trapping would occur.

The database on hydrogen solubility in tungsten is also limited. The results of the two experimental
studies are shown in Figure 6. In the same experi­ments used to determine the diffusivity, Frauenfelder53 also measured solubility. Over the temperature range 1100-2400 K, samples were saturated at fixed pressures and then heated to drive out all of the hydrogen. Over a more limited temperature range of 1900-2400 K, Mazayev et a/.56 also examined hydrogen solubility in tungsten. The agreement with the Frauenfelder’s53 data is quite good in magnitude, but not good in apparent activation energy. As with his diffasivity, the solubility reported by Frauenfelder is typically the value used in predicting the migration of hydrogen in tungsten.

Hydrogen trapping in tungsten has been studied by several research groups. van Veen et a/.57 used bom­bardment by 2 keV protons in their study of the bond­ing of hydrogen to voids in single-crystal tungsten. Thermal desorption from the samples with appms of voids revealed a broad release peak at 600-700 K. It was stated that the release could be modeled as gas going back into solution from the voids with a trap binding energy of 96.5-135 kJ mol-1 controlling the process. Eleveld and van Veen,58 in a similar study, used a lower fluence of 30 keV D+ ions in desorption experiments. In these samples containing vacancies but no voids, the release occurred at 500-550 K. The authors reported a value of 100 kJ mol-1 for the trap binding energy of vacancies. Pisarev eta/.59 used lower fluences of 7.5 keV deuterons into 99.94% pure tungsten samples. During thermal desorption ramps, peaks in the release rates were seen at 350,480,600, and 750 K. The release at the highest temperature was seen only in the highest

fluences. Garcia-Rosales eta/60 used 100 eV deuterium implantation to study the trapping and release rate of hydrogen isotopes from wrought and plasma-sprayed tungsten. Two broad desorption peaks at 475-612 K and 670-850 Kwere seen in the thermal desorption spectra. Modeling of the release data suggested the lower temperature peak to be controlled by both diffu­sion and trapping at a binding energy of 44 kJ mol — . The second release peak was reported to correspond to trapping at defects with a binding energy of 97 kJ mol-1. In experiments with 99.99% pure tungsten and tungsten with 1% lanthanum oxide, Causey et a/.61 examined tritium retention in plasma-exposed samples. Modeling of the results suggested two traps, one with a binding energy of 97 kJ mol-1 and another with 204 kJ mol-1. The density of the trapped tritium averaged 400-500 appm. Anderl et a/.62 used deuterium im­plantation into polycrystalline tungsten to determine the correlation between dislocation density on cell walls and deuterium trapping. Annealing tungsten at 1673 K reduced the dislocation density by a factor of 7, subsequently reducing the deuterium trapping by a similar factor. The binding energy of these traps was estimated to be 88-107 kJ mol-1. As-received 99.95% pure tungsten was used by Sze et a/.63 in experiments with intense deuterium plasma exposure. Exposure at 400 K resulted in blisters with diameters of tens of microns. Elevating the temperature to 1250 K elimi­nated the blisters. Venhaus et a/.64 used high-purity foils in experiments to examine the effect of annealing temperature on blistering by deuterium plasma expo­sure. An unannealed sample and one annealed at 1473

K both exhibited blisters after the plasma exposure. The sample annealed at 1273 K did not blister. There have been a multitude of other reports on blister for­mation on tungsten samples exposed to various forms of hydrogen implantation.65-68

Anderl eta/.62 used 99.95% tungsten in 3 keV D+ ion implantation to determine the recombination — rate coefficient. Over a temperature range of 690­825 K, the recombination rate coefficient was given as kr = 3.85 x 109 exp(-13 500/T)m4 s-1 per mol of H2. This expression is shown in Figure 7, where it is plotted along with the expression given by the Baskes24 model. It can be seen that there is very little correlation between the measured Anderl value and the calculated Baskes value. This is not entirely unusual. Impurities on the surface, especially oxide layers, can have a very strong effect on this coefficient.

While tungsten has excellent low permeability for gaseous tritium, it will be used only in fusion devices as a plasma-facing material. As a plasma­facing material, tungsten will be exposed to intense fluxes of energetic tritium and deuterium. With traps for hydrogen at binding energies of 97 and 203 kJ mol-1(57-62) at natural and radiation-induced defects, it would appear that a substantial tritium inventory could be generated in divertor tungsten. There are several reasons why this high inventory is not likely to occur. The first reason is the high recombination — rate coefficient given earlier. For a recombination — rate constant of 10-1 m4 s-1 per mol of H2 or higher (see Figure 7), the recombination rate on the surface is so rapid as to generate the equivalent of c = 0 at

the boundary. With the very limited penetration dis­tance of energetic hydrogen in the dense tungsten, most of the implanted material is immediately released back out of the surface. There are also recent reports suggesting that ruptured blisters and very fine cracks near the surface69-71 will even further reduce the inward migration of deuterium and tritium into the tungsten.

The influence of neutron irradiation on the thermo­physical properties is related to the irradiation tem­perature and the number of defects generated in the crystal structure. At temperatures <1000 ° C, the electrical232-234 and thermal conductivity217 of tung­sten and tungsten alloys decrease with increasing irra­diation dose. However, at elevated temperatures such as those occurring in a fusion environment, the effect of neutron irradiation is strongly mitigated by anneal — ing.107 Complete recovery of defect-induced material degradation should occur at temperatures >1200 °C (see Figure 8).

In addition to defect generation, material degrada­tion is also related to the formation of transmutation products such as Re and Os, which in general exhibit poorer thermophysical properties. Transmutation — induced degradation increases with increasing tem­perature and irradiation dose, which makes it the most relevant process for the degradation of material properties for future fusion reactors such as DEMO.

Despite the potential for full recovery of the mate­rial defects mentioned above, void-induced swelling occurs. The results235,236 of tungsten and tungsten alloys show that the material’s volume increases with increasing irradiation temperature (<1050 °C).237 W-Re alloys exhibit significantly improved swelling behavior compared to pure W, with a local maxi­mum at ^750 °C. However, the swelling only amounts to <1.7% at 9.5 dpa.237 Therefore, a negligi­ble effect of swelling can be expected for the opera­tion of ITER. Experimental values do not exist at temperatures >1050 °C as expected for the operation of DEMO.

This section describes the present understanding of PWIs for beryllium-containing surfaces. First, it focuses on the erosion properties of ‘clean’ beryl­lium surfaces at different temperatures. Retention of plasma fuel species in both bulk and codeposited layers of beryllium is then described. As beryllium will not be used as the exclusive plasma-facing mate­rial in future confinement devices, issues associated with mixed, beryllium-containing surfaces are also addressed.

4.19.3.1 Beryllium Erosion Properties

The term erosion is used to describe a group of processes that remove material from a surface sub­jected to energetic particle bombardment. Included under the general classification of erosion are pro­cesses such as physical sputtering, chemically assisted physical sputtering, chemical sputtering, and thermally activated release from surfaces. Of these processes, only chemical sputtering, where volatile molecular species are formed on the surface, appears to be inactive in beryllium.

SFT Stacking fault tetrahedral TCH Tension and compression hold

4.20.1 Introduction

Copper alloys are prime candidates for high heat flux applications in fusion energy systems. High heat flux is a major challenge for various fusion devices because of the extremely high energy density required in controlled thermonuclear fusion. The removal of a large amount of heat generated in the plasma through the first wall structure imposes a major constraint on the component design life. Materials with high con­ductivity are needed to assist heat transfer to the coolant and to reduce the thermal stress for pulsed mode of operation.

A number of issues must be considered in the selection of materials for high heat flux applications in fusion reactors. While high conductivity is the key property for such applications, high strength and radiation resistance are also essential for the effective performance of materials in a high heat flux, high irradiation environment. In addition, fatigue behavior is a major concern for many high heat flux applica­tions because of planned or inadvertent changes in the thermal loading. Pure copper has high thermal con­ductivity but rather low strength, and therefore its application as heat sinks is limited. The strength of copper can be improved by various strengthening mechanisms. Among them, precipitation hardening and dispersion strengthening are the two most viable mechanisms for improving the strength of copper while retaining its high electrical and thermal con­ductivities. A number ofprecipitation-hardened (PH) and dispersion-strengthened (DS) copper alloys are commercially available, and have been evaluated for fusion applications, for example, PH CuCrZr, CuNiBe, CuNiSi, and DS GlidCop® Al15, Al25, Al60, MAGT-0.2, etc. Two copper alloys that are most appealing are PH CuCrZr and DS CuAl25. Surveys of copper alloys for fusion applications were conducted by Butterworth and Forty1 and Zinkle and Fabritsiev.2

In this chapter, a brief description of pure copper and several copper alloys of interest to fusion appli­cations is presented, followed by a summary of their physical and mechanical properties. The radiation effects on the physical and mechanical properties of copper and copper alloys as well as their irradiated microstructure are then discussed. Joining techniques for plasma facing components in fusion reactors are also discussed.